Influence of solidification structures on radiation-induced swelling in an additively-manufactured austenitic stainless steel

Bellefon, G. Meric de; Bertsch, K.M.; Chancey, M.; Wang, Y.Q.; Thoma, Dan J.

doi:10.1016/j.jnucmat.2019.06.012

Cited by 45 publications

(9 citation statements)

References 38 publications

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“…In a recent study, it was proved that nanoinclusions inhibited radiation-induced swelling while dislocations promoted it. [50] Therefore, the addition of extra nanoinclusions is expected to increase the anti-swelling property of AM stainless steel. Nevertheless, another study concluded that finer oxide nanoinclusions with an average size of 7.5 nm contributed mostly to strength improvement.…”

Section: Typical Necking Mechanism In Slm Ods 316lmentioning

confidence: 99%

Oxide dispersion strengthened stainless steel 316L with superior strength and ductility by selective laser melting

Zhong

Liu

Zou

et al. 2020

Journal of Materials Science & Technology

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Section: Typical Necking Mechanism In Slm Ods 316lmentioning

confidence: 99%

Oxide dispersion strengthened stainless steel 316L with superior strength and ductility by selective laser melting

Zhong

Liu

Zou

et al. 2020

Journal of Materials Science & Technology

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“…Neutron or ion irradiation could also significantly affect the behaviors (i.e., shortened creep life and reduced fracture toughness) observed in AM alloys [11,[23][24][25].…”

Section: Introductionmentioning

confidence: 99%

Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation (FY21)

Collins¹,

Coq²,

Linton³

et al. 2021

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This report presents the materials property data of additively manufactured (AM) 316L stainless steel (SS) accumulated for the assessment of core materials in the Transformational Challenge Reactor (TCR) program. The TCR manufacturing approach includes using the laser powder bed fusion (LPBF) method for metallic (316L and Inconel 718) components. To assess the mechanical performance of printed components in reactor-relevant conditions and build a property database for the AM materials, mechanical tests and evaluations were performed before and after neutron irradiation. Miniature tensile specimens were irradiated in the High Flux Isotope Reactor to 0.2, 2, 8, and 10 dpa at target temperatures of 300 and 600°C. Postirradiation evaluation for the 0.2 and 2 dpa specimens was performed during FY20 and FY21, and the results are presented and discussed in this document.To obtain the baseline mechanical property data, uniaxial tension testing over a wide temperature range from room temperature (RT) to 600°C was performed for the same materials that were irradiated, including AM 316L SS in the as-built, stress-relieved, and solution-annealed conditions, as well as reference wrought (WT) 316L SS. Regardless of the postbuild heat treatment, the AM 316L SS showed higher strength than the WT 316L SS, but they had similar ductility. A statistical treatment of RT tensile data indicated that variations in the strength and ductility datasets of AM 316L steels were smaller than or similar to those of WT 316L SS. Additionally, 2D mapping of tensile properties for the stress-relieved AM plates also showed clear location dependence of the properties but with limited magnitudes.Postirradiation tensile testing was conducted at RT, 300, 500, and 600°C for selected irradiation conditions. Additional tests were performed near the measured irradiation temperatures of 260 and 390°C for checking the impact of the difference between the targeted and actual irradiation temperatures. Neutron irradiation induced significant changes in the mechanical behavior of the AM SSs, including hardening and softening. Although the as-built 316L SS tested at 300°C after irradiation to 2 dpa at 390°C showed unstable plastic deformation (i.e., necking) immediately after yielding, the overall property changes of the as-printed alloy became much less significant after higher temperature (610 and 690°C) irradiations or when tested at different temperatures. Irradiation-induced ductilization was also observed in the higher strength materials (i.e., in as-built and stress-relieved conditions) after higher temperature (>600°C) irradiations. The strength change after irradiation was generally smaller in the relatively stronger materials (i.e., the as-built and stress-relieved AM SSs) than in the solution-annealed AM and WT SSs. Overall, these relatively lower strength 316L SSs retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after higher temperature irradiations. For the AM 316L mate...

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“…Furthermore, the fracture toughness of additively manufactured materials could be reduced by the increased porosity from the build process, structural anisotropy relative to the build direction, and inclusions from impurities in the feedstock powder [20,30]. Neutron or ion irradiation could also significantly affect the behaviors (i.e., shortened creep life and reduced fracture toughness) observed in additively manufactured alloys [20,31,32,33]. Some of these property degradation mechanisms can be more pronounced in ferritic alloys and ferritic-martensitic steels [34,35].…”

Section: Additive Manufacturing Oxide Dispersion-strengthened Alloysmentioning

confidence: 99%

Feasibility Studies and Downselection of New Materials and Manufacturing Technologies for Nuclear Applications

Byun,

Collins,

Lin

et al. 2023

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Influence of solidification structures on radiation-induced swelling in an additively-manufactured austenitic stainless steel

Cited by 45 publications

References 38 publications

Oxide dispersion strengthened stainless steel 316L with superior strength and ductility by selective laser melting

Oxide dispersion strengthened stainless steel 316L with superior strength and ductility by selective laser melting

Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation (FY21)

Feasibility Studies and Downselection of New Materials and Manufacturing Technologies for Nuclear Applications

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